A conventional boiling water reactor is shown in FIG. 1. Feedwater is admitted into a reactor pressure vessel (RPV) 10 via a feedwater inlet 12 and a feedwater sparger 14, which is a ring-shaped pipe having suitable apertures for circumferentially distributing the feedwater inside the RPV. A core spray inlet 11 supplies water to a core spray sparger 15 via core spray line 13. The feedwater from feedwater sparger 14 flows downwardly through the downcomer annulus 16, which is an annular region between RPV 10 and core shroud 18. Core shroud 18 is a stainless steel cylinder surrounding the core 20, which is made up of a plurality of fuel bundle assemblies 22 (only two 2.times.2 arrays of which are shown in FIG. 1). Each array of fuel bundle assemblies is supported at the top by a top guide 19 and at the bottom by a core plate 21. The core top guide provides lateral support for the top of the fuel assemblies and maintains the correct fuel channel spacing to permit control rod insertion.
The water flows through downcomer annulus 16 to the core lower plenum 24. The water subsequently enters the fuel assemblies 22, wherein a boiling boundary layer is established. A mixture of water and steam enters core upper plenum 26 under shroud head 28. Core upper plenum 26 provides standoff between the steam-water mixture exiting core 20 and entering vertical standpipes 30. The standpipes are disposed atop shroud head 28 and in fluid communication with core upper plenum 26.
The steam-water mixture flows through standpipes 30 and enters steam separators 32, which are of the axial-flow centrifugal type. The separated liquid water then mixes with feedwater in the mixing plenum 33, which mixture then returns to the core via the downcomer annulus. The steam passes through steam dryers 34 and enters steam dome 36. The steam is withdrawn from the RPV via steam outlet 38.
The BWR also includes a coolant recirculation system which provides the forced convection flow through the core necessary to attain the required power density. A portion of the water is sucked from the lower end of the downcomer annulus 16 via recirculation water outlet 43 and forced by a centrifugal recirculation pump (not shown) into jet pump assemblies 42 (only one of which is shown) via recirculation water inlets 45. The BWR has two recirculation pumps, each of which provides the driving flow for a plurality of jet pump assemblies. The pressurized driving water is supplied to each jet pump nozzle 44 via an inlet riser 47, an elbow 48 and an inlet mixer 46 in flow sequence. A typical BWR has 16 to 24 inlet mixers. The jet pump assemblies are circumferentially distributed around the core shroud 18.
The core shroud 18 (see FIG. 2) comprises a shroud head flange 18a for supporting the shroud head 28; a circular cylindrical upper shroud wall 18b having a top end welded to shroud head flange 18a; an annular top guide support ring 18c welded to the bottom end of upper shroud wall 18b; a circular cylindrical middle shroud wall 18d having a top end welded to top guide support ring 18c; and an annular core plate support ring 18e welded to the bottom end of middle shroud wall 18d and to the top end of a lower shroud wall 18f. The diameter of upper shroud wall 18b is greater than the diameter of middle shroud wall 18d, which is in turn greater than the diameter of the lower shroud wall 18f. The entire shroud is supported by shroud support 51, which is welded to the bottom of lower shroud wall 18f, and by annular shroud support plate 52, which is welded at its inner diameter to shroud support 51 and at its outer diameter to RPV 10.
In the event of a seismic disturbance, it is conceivable that the ground motion will be translated into lateral deflection relative to the reactor pressure vessel of those portions of the shroud located at elevations above the shroud support plate 52. Such deflections would cause undesirable stresses on the shroud and its weldments. In particular, if the shroud weld zones have failed due to stress corrosion cracking, there is the risk of misalignment and damage to the core and the control rod components, which would adversely affect control rod insertion and safe shutdown.
Stress corrosion cracking (SCC) is a known phenomenon occurring in reactor components, such as structural members, piping, fasteners, and welds, exposed to high-temperature water. The reactor components are subject to a variety of stresses associated with, e.g., differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other sources such as residual stress from welding, cold working and other asymmetric metal treatments. In addition, water chemistry, welding, heat treatment, and radiation can increase the susceptibility of metal in a component to SCC.
Stress corrosion cracking has been found in the top guide support ring of a BWR. The cracks in top guide support ring 18c extend radially inwardly from the outer circumferential surface thereof and radially outwardly from the inner circumferential surface thereof in the vicinity of the welds which join top guide support ring 18c to shroud walls 18b and 18d.
Stress corrosion cracking in top guide support ring 18c or any other heat affected zone of the shroud girth seam welds diminishes the structural integrity of shroud 18, which vertically and horizontally supports core top guide 19 and shroud head 28. Thus, there is a need for a method and an apparatus for stabilizing a core shroud which has cracks in the weld zones. In particular, there is a need to augment the lateral structural integrity of the shroud for the unlikely case where any of the shroud girth seam welds or heat affected zones thereof are assumed to have failed completely.